1. Field of the Invention
This invention relates to an incore monitoring method and incore monitoring equipment for calculating thermal characteristics of a nuclear reactor continuously, effectively and accurately, and thereby control of a reactor core flow rate, and operation of control rods can be performed adequately based on the calculation result of the thermal characteristics.
2. Description of the Related Art
In adjusting control rod patterns of a boiling water reactor in case of starting-up of the reactor or adjusting reactivity of a reactor core, adjustment is made as a reactor core flow is increased or decreased, and control rods are pulled out or inserted. Since thermal characteristics of a nuclear reactor, such as a maximum linear heat generation rate (MLHGR) or a minimum critical power ratio (MCPR), change every moment in the operation, the operation must be carried out as confirming an output distribution of the thermal characteristics calculated at fixed time intervals or in respect to request from an operator satisfying a condition of an operation limit.
FIG. 14 shows a conventional method of adjusting control rod patterns. That is, in this process, from the start of operation, control rod operation, reactor core flow rate operation, stop of the operation of the control rods and a reactor core flow rate, calculation of power distribution and check of the thermal characteristics are performed in sequence and repeatedly if necessary, and the output is compared with a target output.
In the calculation of the output distribution, thermal characteristics of each fuel assembly composed of the reactor core is calculated based on a nuclear instrumentation system which monitors the neutron flux in a nuclear reactor and the actual measurement of plant data such as the nuclear reactor pressure, the control rod patterns and the reactor core flow rate. The calculation is done by means of a three-dimensional reactor core simulator which combines nuclear calculations for calculating the behavior of neutrons and thermo-hydraulics calculations for calculating the flow distribution inside a reactor core and the void fraction distribution. Consequently, it takes much time to calculate the output distribution, and check the thermal characteristics calculated in the power distribution calculation, and it is thus necessary to suitably interrupt operation during adjusting of the reactor core flow rate or operating of the control rods.
Conventionally, thermal characteristics of sixteen fuel assemblies surrounded by four strings of local power range monitors (LPRMs), which monitor local neutron flux level inside a nuclear core, has been estimated based on the indicated values of LPRM of a detector assembly contained inside four strings. Hereafter, position of a detector assembly is defined by the position of its string.
A conventional calculation of the thermal characteristics based on the indicated value of LPRM is explained with reference to FIG. 15. In this method, the minimum value of the critical power ratios of the sixteen fuel assemblies A1 through A16 surrounded by four strings is calculated, and at each point in height, the maximum value of linear heat generation rates of the sixteen fuel assemblies, are computed by a proportional calculation based on changing rates of values indicated by LPRMs of detector assemblies B1 through B4 each belonging to one of four strings.
In the conventional method of core monitoring, if the severest fuel assembly in terms of the thermal characteristics is not one of fuel assemblies A1, A4, A13 and A16 in FIG. 15, which are adjacent to a string, the LPRM detectors of detector assemblies B1 through B4 are away from (not adjacent to) the severest fuel assembly in terms of the thermal characteristics. Moreover, if the severest fuel assembly in terms of the thermal characteristics is one of the fuel assemblies A1, A4, A13, and A16, the values indicated by the LPRM detectors, one of which is close to but the others are away from the severest fuel assembly, are used in the computation. That is, for example, if the severest in terms of the thermal characteristics is a fuel assembly A1, detector assemblies B2 through B4 are away from the fuel assembly A1, and thus the thermal characteristics of the fuel assembly A1 are less correlated with values indicated by the LPRM detectors of the detector assemblies B2 through B4, and it is difficult to compute the thermal characteristics with a high degree of accuracy by the conventional method.
As mentioned above, in order to monitor the thermal characteristics at the time of operating of the reactor core flow rate or control rods in a boiling water reactor, the calculation takes a long time, and the operation of controlling the control rods and adjusting the reactor core flow rate must be stopped in at each cycle through FIG. 14. And when evaluating the thermal characteristics easily computed based on values indicated by LPRM monitors, the thermal characteristics cannot be evaluated with sufficient accuracy, thus it is necessary to allow a big margin for arrangement of core soundness.
Accordingly, it is an object of embodiments of this invention to provide an incore monitoring method which is able to compute thermal characteristics of a nuclear reactor more rapidly and continuously with a high degree of accuracy, and thereby to maintain fuel soundness and to shorten the time necessary for starting-up or adjusting fuel arrangement patterns.
Other and further objects of this invention will become apparent upon an understanding of the illustrative embodiments to be described herein or will be indicated in the appended claims while various advantages not referred to herein will be apparent to one skilled in the art upon employment of the invention in practice.
According to one aspect of the invention, there is provided an incore monitoring method of a nuclear reactor, including, measuring neutron flux levels at a plurality of pitch levels corresponding to a plurality of local power range monitor sensors arranged in an axial direction inside a detector assembly installed in the nuclear reactor, performing a power distribution calculation using a three-dimensional simulation to obtain a first calculation of thermal characteristics of a fuel assembly group consisting of fuel assemblies adjacent to the corresponding detector assembly, based on values indicated by the plurality of local power range monitor sensors of the corresponding detector assembly at a first time, performing a plurality of second calculation of thermal characteristics in which the power distribution calculation is not performed, based on values indicated by the plurality of local power range monitor sensors at second times, subsequent to the first time, and based on the thermal characteristics calculated in the first calculation at the first time, and monitoring the plurality of thermal characteristics calculated in the second calculation.
According to one aspect of the invention, there is provided an incore monitoring equipment of a nuclear reactor, including, a detector assembly configured to be installed in the nuclear reactor, including local power range monitor sensors to measure neutron flux levels at a plurality of pitch levels, a three-dimensional simulator for calculating a power distribution including first calculation of thermal characteristics of a fuel assembly group consisting of fuel assemblies adjacent to the corresponding detector assembly, based on values indicated by the plurality of local power range monitor sensors of the corresponding detector assembly at a first time, and a monitoring unit for performing a plurality of second calculations of thermal characteristics in which the power distribution calculation is not performed, based on values indicated by the plurality of local power range monitor sensors at second times, subsequent to the first time, and based on the first calculated thermal characteristics at the first time.